Counter Current Flow Limitation (CCFL) was observed in the pressurizer surge line of the Oregon State University APEX facility during test NRC-10. This test simulated a one-inch diameter cold leg break with a failure of three of four of the fourth-stage Automatic Depressurization System (ADS) valves. The result was a high vapor flow rate through ADS 1-3, that caused CCFL in the pressurizer surge line and liquid holdup in the pressurizer. Because this liquid was not available for core cooling, further study of the passive safety systems in the AP600 under Mode 5 Cold Shutdown conditions was deemed necessary. An analysis of the AP600 geometry and the existing CCFL database determined that Kutateladze scaling is appropriate for the APEX and AP600 surge lines. The Kutateladze CCFL correlation was used to assess CCFL in the APEX and AP600 pressurizer surge lines under Mode 5 Cold Shutdown conditions. The results indicate that CCFL would be expected in the pressurizer surge lines at low pressures and decay powers prior to ADS 4 actuation. Test NRC-35 examined CCFL and provided data to benchmark NRC's thermal hydraulic analysis codes. This thesis presents the results of test NRC-35 and the supporting CCFL calculations. / Graduation date: 1999

When soluble boron is used to control reactivity, there are two particular events which can challenge long-term core cooling (LTCC) during the small break loss-of-coolant accident (SBLOCA): boron precipitation and boron dilution. The initial consequences of the SBLOCA are mitigated by the emergency safety systems, but the core continues to boil. As boron is less volatile than steam, the steam is virtually boron free. All the boron remains in the core, the boron concentration in the core rises. If the solubility limit is reached, precipitation could occur. The boron precipitation event was historically considered to be bounded by the large break accident. However, there are characteristics of the SBLOCA which cannot be neglected and an SBLOCA specific methodology is required. On the opposite end of the boron concentration spectrum is the SBLOCA boron dilution event. The steam generators remove heat from the primary and condense the steam. The condensation of the boron-free steam can result in the accumulation of a deborated slug of water. If natural circulation restarts, the slug can be transported toward the core and potentially reduce the core boron concentration enough to induce a recriticality. This thesis describes two analytical methodologies for these SBLOCA LTCC events. The two methodologies have a similar approach. Both use transient system analyses for inputs to and justification of the follow-on boron concentration calculations. For boron precipitation, a maximized concentration is calculated with the Small Break Boron Precipitation model. For boron dilution, a minimized core inlet concentration is calculated using computational fluid dynamics. / Master of Science

Die Veranstaltung widmete sich mit der Borverdünnung in Druckwasserreaktoren bzw. mit der Verstopfung der Sumpfansaugsiebe durch freigesetztes Isolationsmaterial schwerpunktmäßig zwei Themen der Reaktorsicherheit, die auch in aktuellen Aufsichtsverfahren eine Rolle spielen. Eingebettet in den internationalen Kontext wollten die Veranstalter die sicherheitstechnische Bedeutung dieser Themen für die deutschen Anlagen beleuchten und die Auswirkungen auf die zu erbringenden Sicherheitsnachweise und den Anlagenbetrieb darstellen. Dabei kamen Gutachter, Vertreter der Forschung, Hersteller und Betreiber gleichermaßen zu Wort. Der Fachtag sollte den Teilnehmern aber insbesondere vermitteln, welche Beiträge die privat und öffentlich finanzierte Reaktorsicherheitsforschung zur Aufklärung der jeweiligen Ereignisabläufe und ihrer sicherheitstechnischen Bedeutung geleistet hat. In diesem Forschungskontext spielen, auch international, die Methoden der so genannten Computational Fluid Dynamics (CFD) eine zunehmende Rolle. Deshalb widmete sich eine Sitzung den Grundlagen, Möglichkeiten und Grenzen von CFD-Methoden. Dabei wurden u.a. Anwendungen zur Borvermischung und zum Verhalten von Mineralwolle im Sumpf präsentiert.

Die Veranstaltung widmete sich mit der Borverdünnung in Druckwasserreaktoren bzw. mit der Verstopfung der Sumpfansaugsiebe durch freigesetztes Isolationsmaterial schwerpunktmäßig zwei Themen der Reaktorsicherheit, die auch in aktuellen Aufsichtsverfahren eine Rolle spielen. Eingebettet in den internationalen Kontext wollten die Veranstalter die sicherheitstechnische Bedeutung dieser Themen für die deutschen Anlagen beleuchten und die Auswirkungen auf die zu erbringenden Sicherheitsnachweise und den Anlagenbetrieb darstellen. Dabei kamen Gutachter, Vertreter der Forschung, Hersteller und Betreiber gleichermaßen zu Wort. Der Fachtag sollte den Teilnehmern aber insbesondere vermitteln, welche Beiträge die privat und öffentlich finanzierte Reaktorsicherheitsforschung zur Aufklärung der jeweiligen Ereignisabläufe und ihrer sicherheitstechnischen Bedeutung geleistet hat. In diesem Forschungskontext spielen, auch international, die Methoden der so genannten Computational Fluid Dynamics (CFD) eine zunehmende Rolle. Deshalb widmete sich eine Sitzung den Grundlagen, Möglichkeiten und Grenzen von CFD-Methoden. Dabei wurden u.a. Anwendungen zur Borvermischung und zum Verhalten von Mineralwolle im Sumpf präsentiert.

Examensarbetet behandlar problematik som uppstår vid härdsmältningsförlopp i en kärnkraftsreaktor av typen kokvattensreaktor. Resultatet ska användas som riktlinjer till strategier som utvecklas av Severe Accident Management Guidelines (SAMG) där arbetets uppdrag är ett delmoment i framtagning av strategier för att bemästra de olika fenomen som uppstår vid härdsmälta.   Syftet med arbetet är att undersöka begränsningar för att bevara reaktortanken intakt vid haveri, genom att undersöka den minsta mängd vatten som behövs för att undvika tankgenomsmältning. Undersöka fallen som leder till härdsmälta och gruppera dem efter händelsesekvenser. Undersöka metall/vatten-reaktionen som uppstår då härden blir över 800°C och undersök om fallen kan grupperas i händelsesekvenser.  Metoder som används i arbetet är PSA-dokumentation, händelseutvecklingsträd, teoretiska beräkningar och MAAPv5.03. Resultatet beskriver att grupperingar av fallen som slutar i härdsmälta och grupperingar av metall/vatten-reaktionen hos de olika fallen kan genomföras. Resultatet beskriver även ett minsta flöde som kan föras in i reaktortanken för att hindra tankgenomsmältning och flöden upp till 100 kg/s så det finns resultat för olika flöden beroende på vilka kylmedel som är tillgängliga.  Slutsatsen av arbetet är att fall kan grupperas efter händelsesekvenser och påverkan hos metall/vattenreaktion, grupperingarna sparar tid vid ett haveriförlopp. I varje grupp kunde det svåraste fallet beräknas för minsta flöde för att klara tankgenomsmältning och flöden upp till 100 kg/s.  Framtida arbeten bör undersöka trycket och vätgasen som skapas vid vatteninmatning samt dess påverkan på reaktorinneslutningen. / The master thesis deals with problems that arise during nuclear meltdown in a nuclear powerplant of the type boiling water reactor. The work will be used as guidelines for strategies developed by Severe Accident Management Guidelines (SAMG), this master thesis is a sub-element in the development of strategies for mastering the various phenomena that arise during a meltdown.  The purpose of the work is to investigate limitations for maintaining the reactor tank intact during the meltdown by, examining the minimum amount of water needed to avoid the meltdown getting through the reactor tank. Examining the cases that lead to meltdown and group them according to the event sequences. Examine the metal/water-reaction that occurs when the core becomes over 800°C and examine if the cases can be grouped into event sequences.  Methods used in the master thesis is PSA-analysis, event development threes, theoretical calculations and MAAPv5.03.  The result describes the groupings of the cases ending in meltdown and the groupings of the metal/water-reaction of the various cases. The result also describes a minimum flow that is required to prevent meltdown of getting through the reactor tank and flow up to a 100 kg/s.  The conclusion of the master thesis is that cases can be grouped according to event sequences and the influence of the metal/water-reaction, the groupings save time in the event of a breakdown. In each group the most difficult case was calculated so that the lowest flow to prevent the meltdown from getting through the reactor tank was presented among with different flows up to 100 kg/s. Future work should investigate the pressure and hydrogen gas created by the water input and its influence on the reactor inclusion.

This master's thesis is an investigation and evaluation of MELCOR (a software tool for severe accident analyses regarding nuclear power plants), or more correctly of the (ASEA-Atom BWR 75) reactor model developed for version 1.8.6 of MELCOR. The main objective was to determine if MELCOR, with the reactor model in question, is able to produce satisfactory results in severe accident analyses compared to results made by MAAP, which is currently the only official software tool for this application in Sweden. The thesis work is related to the RASTEP project. This project has been carried out in several stages on behalf of SSM since 2009, with a number of specific issues explored within an NKS funded R&D project carried out 2011-2013. This investigation is related to the NKS part of the project. The purpose with the RASTEP project is to develop a method for rapid source term prediction that could aid the authorities in decision making during a severe accident in a nuclear power plant. A software tool, which also gave the project its name, i.e. RASTEP (RApid Source TErm Prediction), is therefore currently under development at Lloyd's Register Consulting. A software tool for severe accident analyses is needed to calculate the source terms which are the end result from the predictions made by RASTEP. A set of issues have been outlined in an earlier comparison between MAAP and MELCOR. The first objective was therefore to resolve these pre-discovered issues, but also to address new issues, should they occur. The existing MELCOR reactor model also had to be further developed through the inclusion of various safety systems, since these systems are required for certain types of scenarios. Subsequently, a set of scenarios was simulated to draw conclusions from the additions made to the reactor model. Most of the issues (pre-discovered as well as new ones) could be resolved. However the work also rendered a set of issues which are in need of further attention and investigation. The overall conclusion is that MELCOR is indeed a promising alternative for severe accident analyses in the Swedish work with nuclear safety. Several potential benefits from making use of MELCOR besides MAAP have been identified. In conclusion, they would be valuable assets to each other, e.g. since deviations in the results (between the two codes) would highlight possible weaknesses of the simulations. Finally it is recommended that the work on improving the MELCOR reactor model should continue. / RASTEP